Reprocessing of spent uranium fuel by means of chlorination and fractional absorption by barium chloride



United States 3,372,998 REPRGCESSTNG F SPENT URANIUM FUEL BY MEANS OF CHLORINATXON AND FRACTION- AL ABSORPTION BY BARIUM CHLORIDE Tairehiko Ishihara and Kenmei Hirano, Mito-shi, Ibarakiken, and Takeshi Soga, Naka-gun, Ibaraki-ken, Japan, assigners to Japan Atomic Energy Research Institute, Tokyo, Japan No Drawing. Filed Mar. 4, 1966, Ser. No. 531,773 Claims priority, application Japan, Mar. 18, 1965, 40/ 15,390 Claims. ((13. 23-326) ABSTRACT OF THE DISCLOSURE This invention relates to the reprocessing of spent nuclear fuel and particularly to a process for separating fission products from a spent uranium fuel.

Extraction of the spent fuel with organic solvents is the only reprocessing method now in practical use. This method has a very high decontamination factor, but the spent fuel must be stored or cooled prior to reprocessing for about 100 days until its radioactivity decreases substantially to avoid decomposition of the organic solvent by radiation. According to another known method, the spent fuel is fluorinated and the vapors of the fluorides obtained are passed through a bed of solid sodium fluoride at 100 C., whereby uranium hexafluoride is selectively absorbed by the salt. When the salt bed is heated to 400500 C., uranium fluoride is selectively desorbed from the salt. By repeating the absorption and desorption cycle a decontamination factor of 1x 10 is reached. However, fluorine and the vapors of fluorides are highly corrosive, making it extremely difiicult to construct an industrial reprocessing plant.

The fractional distillation of chlorides obtained by chlorinating spent uranium fuel has been studied. But the decontamination factor of the known fractionation methodsis too low.

The object of this invention is a more efficient method for fractionating chlorinated spent uranium fuel.

According to this invention, spent uranium fuel is reacted with a chlorinating agent that is a carbon compound having no hydrogen in its molecule to form a mixture of uranium chlorides, mainly uranium tetrachloride, and chlorides of accompanying fission products. The mixed chloride vapors are passed through a bed of solid barium chloride, in which uranium is abof the fission products, whose vapor pressures are high, pass through the bed. The absorbed uranium tetrachloride is further chlorinated by passing chlorine gas through said bed, whereby gaseous uranium hexachloride is released.

The spent uranium may be in the form of metallic uranium or of uranium compounds (oxide, carbide) suitable primary chlorinating agents are carbon tetrachloride, phosgene, hexachloroethane or tetrachloro:

atent ethylene, and are employed at 450-700" at about 600 C.

During primary chlorination, uranium in the fuel ma terial is mainly converted to uranium tetrachloride. A small amount of pentachloride and hexachloride may also be produced. The fission products form chlorides most of which are volatile.

When the vapor mixture of chlorides which is substantially free of chlorine gas passes through a bed of barium chloride granules which is kept at 250-580 C., uranium tetrachloride is believed to react with barium chloride to form a compound Ba UCl which is stable at temperatures below 583 C. Uranium pentachloride and hexachloride in the gaseous mixture are unstable in the absence of a substantial amount of chlorine and are converted to tetrachloride while passing through the barium chloride bed. Depending upon the temperature of the barium chloride bed, chlorides of the rare earths, of zirconium and colurnbium may also be absorbed by the barium chloride.

After the mixture is absorbed, the bed is kept at a temperature between 350 C. and 580 C. (higher temperatures are desirable in this stage) while a gas free from chlorine is being passed in a continuous stream to remove chlorides of the fishion products.

Then the bed is heated to a temperature between 400 C. and 700 C. while chlorine or a mixture of chlorine and an inert gas is being passed through the bed in a continuous stream. The uranium tetrachloride is converted to hexachloride and removed from the bed. The vapor pressure of UCl is 1 atm. at 280 C. At a temperature higher than 700 C., chlorides of the fission products absorbed in the bed are volatilized. Therefore, the desorption should be carried out at a temperature lower than 700 C.

The uranium hexachloride may be reduced to tetrachloride in a chlorine-free atmosphere and absorbed in another barium chloride column for repetition of the separation treatment. Very pure uranium can be recovered.

C., preferably Example Uranium dioxide fuel pellets, which had been irradiated with a total neutron flux of 1X10 n./cm. and then had been cooled for 50 days, were pulverized by alternating oxidation with air and reduction with hydrogen at elevated temperature.

One gram of the impure uranium dioxide powder obtained was heated to 600 C. in a quartz tube of 25 mm. diameter and 20 cm. length, and a mixture of carbon tetrachloride vapor and argon (40:60 by volume) was passed through said tube. The mixed chloride vapor thus produced was passed through a quartz column of 20 mm. diameter and 20 cm. length packed with barium chloride granules (4 mm. in diameter) heated to 350 C.

The barium chloride tube was then cooled to room temperature so as to strengthen the chemical bond between the uranium tetrachloride and the barium chloride. It was again heated to 500 C. while a mixture of carbon tetrachloride vapor and argon (20:80 by volume) was passed therethrough for 30 minutes, whereby the volatile o'hlorides of the fission products were removed as much as possible.

Then a mixture of carbon tetrachloride and chlorine (:15 by volume) was passed through the barium chloride bed at 500 C., whereby the absorbed uranium tetrachloride was oxidized to uranium hexachloride, which was vaporized and deposited on a cool tube connected to the barium chloride tube. The hexachloride was reduced to the tetrachloride by standing in a chlorine-free atmosphere, and the uranium tetrachloride was collected and weighed. Recovery of uranium was 99%. A small mount of the collected uranium tetrachloride was used 'or gamma ray spectrography. The decontamination facor with respect to gamma radiation was 95.

In Experiment I, the effect of varying proportions of CCl, in the chlorinating gas mixture was checked.

One half gram of the collected tetrachloride was heated TABLE I 400 C. in the 40:60 mixture of carbon tetrachloride 1 2 3 Iapor and argon, and the vapor was absorbed on barium :hloride in a tube preheated to 400 C. and kept at 500 Gas Composition 01- A CCl-70A 20 001 -80A 3. for 30 minutes while the gas mixture passed through ggfll gt itgfig 40C 4 60 r 30 4 r r he tube. Then the tube was kept at 500 C. and the 2 30 28 Absorption Time nixture of carbon tetrachloride and chlorine was passed (mm) 35 45 60 ind uranium hexach'loride was recovered. Recovery was z ggg g Uranium 99 6 98 6 99 O 99% of the recycled material or 98% of the initial g gg 'f imount, and the decontamination factor was 3.6 X10 Factor 85 93 78 It was also found that the temperature of the absorption :ube packed with barium chloride should vary in the difection of gas flow, the inlet end being heated to 500 The concentration of CO1. 1n the gas mixture has no suband the temperature f the tube gradually decreasing stantial influence upon the chlorinat1on of the spent ura- :0 150 C. at the outlet end. This permits uranium pentamum flmL L; n e used. However, when :hloride and hexachloride to be reduced to the tetrachlothe concentratlon 1S f than 50 Q p j F 4 15 ride while they pass through the absorption tube. Graded P m and (1690510011 0f caTbQH 15 deposlte'd 1n the heating f absorption tube was therefnre employed reaction tube. In all later experiments, a mixture of in the experiments whose results are summarized in 40 4 Was therefore P Y Gas mlxturas Table used for oxidation and desorption of the absorbed ura- All experiments were carried out as described in the nium tetrachloride mostly contamed argon as a carrier first half'of the above example. The uranium fuel materials used in these experiments had been irradiated and In Experiment 11, 9 argon Was used the U0 samples were pulverized by the above-mentioned chlorides Of the fission Product, and condltlons Were oxidation and reduction cycle. Metallic uranium and otherwise the same as in Experiment I. The results were U0 were directly chlorinated. unsatisfactory.

TABLE 0 Experiment Experimental Conditions I II III IV V VI VII Sample:

COmIJOllIId U02 U01 U02 U01 Irradiation Dose (nJcmfi) 2 10 No. of Cooling Days..- 40 Amount Used (g.) 0.5..-

Chlorination and Absorptio Gas Composition (vol. percent) Table 1..

Gas Flow Rate (cm/min. 25 0.) Amount of Absorbent (g.) Chlorination Temperature C 0 Absorption Temperature 0.) Absorption Time (min.) Selective Desorption oi Fission l? chlorides:

Gas Composition (vol. percent) 10 001i 100 Ar 20 0014 10 C014 5 001 5 0014 10 001 90 Ar 80 Ar 90 Ar 95 Hz 95 Ar 90 N2. Gas Flow Rate (cmJmin. 25 C.) 32 29 40 40 32. Temperature 0.). 4R0 480 450 450.- Table V 480 490. Time (min. 60. 60 30 60. Selective Desorption of Uranium Hexachlonde Gas Composition (vol. pereen 70 Cl: 70 0h 90 011 Table IV 100 Ch 90 01, 80 Ch.

15 0014 15 0014 10 C014 10 0014--- 20 0014. 15 Ar 15 Ar Gas Flow Rate (emJmin. 25 C.) 96. 96 80. do 70 96- 80. Temperature 0.). 540 540-.. 560 530 540 Table VI 540. Time (min) 90... 90 90 9 0 90 90. Recovery of Uranium (percent) Table I.-. 97.0 Table III.

Decontamination Factor (Gamma Radio Table IL.-- o

Experiment Experimental Conditions VIII IX X XI XII XIII XIV Sample:

Compound... U0

Irradiation Dose (nJomJ) 2X10" No. of Cooling Days 50 Amount Used (g.) 0.5 Chlorination and Absorption:

Gas Composition (vol. percent) 40 C014 30 0 01 40 C01 60 Na 60 Ar Ar 0 Ar 60 Ar 60 Ar 0 N2.

Gas Flow Rate (cm/min. 25 C.) i2

Amount of Absorbent (g.) 22

Chlorination Temperature 0.) un

Absorption Temperature 0.) 500-150 Absorption Time (min) Selective Desorption oi Fission Product Chlorides:

Gas Composition (vol. percent) 100 N2 I(i014 5 C012 5 03516 10 001; 40 0014 100 Ni Gas Flow Rate (cm./min. 25 0.). 32.

Temperature 0.) 520;.

Time (min) 60 Selective Desorption of Uranium Hexaehlorldez Gas Composition (vol. percent) 012 70 Cl:

Gas Flow Rate (cmJmin. 25 0.) 80 96-..

Temperature C.) 540 550 Time (min) 9O Recovery of Uranium (percent) 97.3 99.9 98.3 99.1 Decontamination Factor (Gamma Radioactivity) 88 2.0)(10 68 78 93 In Experiment III, the chlorination temperature was varied. The results are shown in Table III.

TABLE III Chlorination Temperature O.) 450 500 600 700 Absorption Time (min) 360 180 G0 45 Recovery of Uranium (percent) 94. 7 98.2 99.4 99.0 Decontamination Factor 2.1)( 1. 2X10 1.3)(10 72 TABLE IV Composition of Desorbing Gas (vol. percent) 80 012- CC]! 90 012-10 C014 1C0 Cl; Recovery of Uranium (percent) 9. 0 99. 2 98. 1 Decontamination FactoL 1. 3x10 1. 5X10 95 In Experiment V, the temperature for desorbing fission product chlorides was varied, and hydrogen Was used as a diluent for CCL; in the desorbi-ng gas mixture. The results are shown in Table V, and indicate TABLE V Fission Products Desorption 350 400 450 500 550 580 Temperature (C Recover oruranium'oifcfiiif 97.5 96.2 98.8 98.0 98.7 93.2 Decontamination Factor 45 53 82.1 1.1 10 97 90 TABLE VI Desorption Tem 400 500 530 540 550 580 600 700 i iffii. Recovery of Ur8n- 45. 0 96. 7 99. 2 99. 8 98. 7 91. 0 78. 5 58. 7

Decontamination 1. 7X10 2 92 85 88 79 67 42 In Experiment VII, nitrogen was tested successfully as a diluent for CCl, in the chlorination of fuel material and in the desorption of fission products.

Experiment VIII differs from Experiment VII, in that only nitrogen was used for desorbing fission product chlorides. As shown in Table 0, 100% nitrogen can be used though not with best results.

Experiment IX was carried out with U 0 obtained by stopping the oxidation-reduction cycle of the Example after the oxidation stage. In this experiment, hydrogen was used as a diluent for CC], in the desorption of the fission products. The results were excellent.

In Experiment X, phosgene was used as a chlorinating agent though not with completely satisfactory results.

In Experiment XI, hexachloroethane was used successfully as a chlorinating agent.

In Experiment XII, metallic uranium was used as the starting material. A one gram pellet of metallic uranium was directly chlorinated without any pretreatment.

In Experiment XIII, uranium dicarbide was used successfully as the starting material for the process.

In Experiment XIV, the procedure of Experiment XIII 6 was repeated with phosgene as a chlorinating agent and nitrogen as a diluent. The results were satisfactory.

In further experiments, the following compounds as chlorinating agents were tested.

CCl CCl Tetrachloroethylene C Cl Hexachlorobenzene CNCl Cyanogen chloride C OCI Hexachlorophenol O C CL, O Tetrachloroquinone C1 C0 CCl Trichloromethyl chloroformate Diphosgene) C N Cl Tricyanogen chloride CCl CCl-CCl CCl Hexachlorol ,B-butadiene The color of the reaction product in the chlorination tube indicated that all these compounds form uranium tetrachloride and are applicable to this process, though the yields have not yet been determined.

The method of this invention has the following advantages:

(1) By repeated absorption and desorption of uranium chloride in a barium chloride bed, uranium can be recovered with a high decantamination factor.

(2) The operation is anhydrous so that there is no corrosion problem.

(3) The fission products are recovered in the solid state, as a non-volatile residue absorbed in the barium chloride bed or as a non-volatile residue condensed after being desorbed from the barium chloride without further treatment.

(4) Uranium can be recovered with a yield of 99% by weight.

(5) The vapor pressure of barium chloride is so low (1X 10 atm. at 1000 C., B.P. 1560 C.) and its melting point is so high (960 C.) that the recovered uranium cannot be contaminated with vaporized barium chloride.

What we claim is:

1. A method of reprocessing spent nuclear fuel essentially consisting of a material selected from the group consisting of uranium metal, uranium oxide, and uranium carbide, and of contaminating fission products of said material, which comprises:

(a) reacting said fuel with a chlorine-bearing carbon compound free from hydrogen at 450 to 700 C. until said material is substantially converted to uranium tetrachloride vapor and said fission products are converted to vapors of the respective chlorides;

(b) contacting said vapors with solid barium chloride at a temperature not substantially lower than C. and lower than 583 C., whereby said uranium tetrachloride and a portion of the chlorides of said fission products are absorbed by said barium chloride;

(0) contacting said barium chloride while having said uranium tetrachloride absorbed therein with elementary chlorine at a temperature substantially between 400 and 700 C. until uranium hexachloride as a vapor is discharged from said barium chloride; and

(d) collecting said uranium hexachloride.

2. A method as set forth in claim 1, wherein said carbon compound consists of carbon and chlorine.

3. A method as set forth in claim 1, wherein said carbon compound consists of carbon, chlorine, and oxygen.

4. A method as set forth in claim 1, wherein said carbon compound consists of chlorine, carbon, and nitrogen.

5. A method as set forth in claim 1, wherein said carbon compound is carbon tetrachloride.

6. A method as set forth in claim 1, wherein said carbon compound is hexachloroethane, tetrachloroethylene, or hexachloro-1,3-butadiene.

7. A method as set forth in claim 1, wherein said carbon compound is phosgene.

8. A method as set forth in claim 1, wherein said carbon compound is gaseous and diluted with argon.

References Cited UNITED STATES PATENTS 3,125,409 3/1964 Tury 23337 3,165,376 1/1965 Golliher 23337 3,178,258 4/1965 Cathers et a1. 23-337 L. DEWAYNE RUTLEDGE, Primary Examiner.

BENJAMIN R. PADGETT, Examiner.

10 S. TRAUB, M. J. McGREAL, Assistant Examiners. 

